• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3225-3234
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• 2022

Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.

• Nuclear Engineering and Technology
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• 제45권1호
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• pp.73-80
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• 2013

The primary water stress corrosion cracking (PWSCC) of Alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM), pressurizer instrumentation, and the pressurizer heater sleeves. Recently, two cases of boric acid precipitation that indicated leaking of the primary cooling water were reported on the bottom head surface of steam generators (SG) in Korea. The PWSCC resistance of Ni base alloys which have intergranular carbides is higher than those which have intragranular carbides. Conversely, in oxidized acidic solutions like sodium sulfate or sodium tetrathionate solutions, the Ni base alloys with a lot of carbides at the grain boundaries and shows less stress corrosion cracking (SCC) resistance. The role of grain boundary carbides in SCC behavior of Ni base alloys was evaluated and effect of intergranular carbides on the SCC susceptibility were reviewed from the literature.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1998년도 특별강연 및 추계학술발표 개요집
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• pp.288-291
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• 1998

Inconel 600 alloy is used as the material of nuclear steam generator tubing because of its mechanical properties, formability, and corrosion properties. According to reports, the life time of nuclear power plants decreases because of the pitting, intergranular attack, primary water stress corrosion cracking(PWSCC), and intergranular stress corrosion cracking(IGSCC), and denting in the steam generator. The SCC test is very important because of SCC appears in various environment such as solutions, materials, and stress. The C-Rig specimen was made of the steam generator welded sleeve repairing by the pulsed Nd:YAG laser. In the corrosion invironment, corrosion solutions are Primary Water, Caustic, and Sulfate solution and corrosion time is 1624-4877hr. The permitted stress is 30-60ksi.In this C-Ring SCC test is the relationship between corrosion depth, crack and corrosion environment is evaluated. SCC was happens in Sulfate and Corrosion solution but doesn't happen in Primary Water. The corrosion time and stress is very affected by the severely environment of Sulfate or Caustic solution. The microstructure observation indicates that SCC causes interganular failure in the grain boundary of vertical direction.

• 대한기계학회논문집A
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• 제36권9호
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• pp.1047-1052
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• 2012

가압경수로형 원전의 Alloy 600 원자로압력용기헤드 관통노즐 및 Alloy 82/182 이종금속 맞대기 용접부에서 일차수응력부식균열(PWSCC)이 보고된 이후 전 세계적으로 PWSCC에 의한 용접부 파단을 예방하기 위해 강화검사를 적용하고 있다. 본 이종금속용접부에 대한 가동중검사에서 균열이 발견된 경우 건전성평가 결과가 도출되기까지 발전소가 정지 상태에 있게 됨에 따라 원전 이용율 저하가 발생할 수 있는데, 이를 예방하기 위해서는 균열건전성평가 관련 기술의 정립뿐만 아니라 신속하게 평가 결과를 도출할 수 있는 시스템의 구축이 필요하다. 본 연구에서는 이종금속 맞대기 용접부를 대상으로 진행하고 있는 PWSCC 균열건전성평가 기준 정립 및 전산 시스템 개발 결과를 제시하였다. 본 연구를 통해 이종금속 맞대기 용접부 PWSCC 균열건전성평가 기술이 정립되고 전산 시스템으로 구현되어 원자로압력용기 주변 이종금속 맞대기 용접부에서의 PWSCC 균열에 대한 기술적 건전성평가 수단을 확보하였다.

• 대한기계학회논문집A
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• 제34권1호
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• pp.97-101
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• 2010

니켈합금 용접재료인 Alloy 82/182 용접부에서의 일차수응력부식균열(PWSCC, Primary Water Stress Corrosion Cracking)은 원자력발전소내 주요 기기의 건전성을 저해시킬 수 있는 요인으로, 용접시 발생하는 인장 잔류응력에 의해 발생할 수 있다. 해외 원자력발전소의 경우 가압기 노즐 등에 적용된 Alloy 82/182 이종금속 용접부에서 PWSCC에 의한 균열이 여러 차례 보고되고 있으며, 이를 예방하기 위한 법으로 인장 잔류응력을 줄여줄 수 있는 오버레이 용접을 수행하고 있다. 본 논문에서는 PWSCC를 예방하기 위한 목적으로 수행되는 오버레이 용접에 대해 설명하고 오버레이 용접의 효과를 유한요소해석을 통해 확인하였다.

• Corrosion Science and Technology
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• 제3권5호
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• pp.198-208
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• 2004

Although there has been no general agreement on the mechanism of primary water stress corrosion cracking (PWSCC) as one of major degradation modes of Ni-base alloys in pressurized water reactors (PWR's), common postulation derived from previous studies is that the damage to the alloy substrate can be related to mass transport characteristics and/or repair properties of overlaid oxide film. Recently, it was shown that the oxide film structure and PWSCC initiation time as well as crack growth rate were systematically varied as a function of dissolved hydrogen concentration in high temperature water, supporting the postulation. In order to understand how the oxide film composition can vary with water chemistry, this study was conducted to characterize oxide films on Alloy 600 by an in-situ Raman spectroscopy. Based on both experimental and thermodynamic prediction results, Ni/NiO thermodynamic equilibrium condition was defined as a function of electrochemical potential and temperature. The results agree well with Attanasio et al.'s data by contact electrical resistance measurements. The anomalously high PWSCC growth rate consistently observed in the vicinity of Ni/NiO equilibrium is then attributed to weak thermodynamic stability of NiO. Redox-induced phase transition between Ni metal and NiO may undermine the integrity of NiO and enhance presumably the percolation of oxidizing environment through the oxide film, especially along grain boundaries. The redox-induced grain boundary oxide degradation mechanism has been postulated and will be tested by using the in-situ Raman facility.

• 한국압력기기공학회 논문집
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• 제8권1호
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• pp.8-18
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• 2012

In early constructed nuclear power plants, Ni-based Alloys 82/182 had been widely used for dissimilar metal welds (DMW) as a weld filler metal. However, Alloys 82/182 have been proven to be susceptible to primary water stress corrosion cracking (PWSCC) in the nuclear primary water environment. The formation of crack due to PWSCC is also influenced by weld residual stresses. Thus, the accurate estimation of weld residual stresses of DMW is crucial to investigate the possibility of PWSCC and instability behaviors of crack due to PWSCC. In this context, the present paper investigates weld residual stresses of nuclear reactor piping nozzles based on 2-D axi-symmetric finite element analyses based on layer-based approach using maximum molten bead temperature. In particular, the effect of analysis parameters, i.e., a thickness of weld layer, an initial molten bead temperature, convection heat transfer coefficient, and geometric constraints on predicted weld residual stresses was investigated.

• 부식과 방식
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• 제12권1호
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• pp.30-38
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• 2013

원전의 구조적 건전성에 문제가 될 수 있는, 오스테나이트계 합금의 환경조장균열(EAC)에 대한 거동을 실험적인 결과와 문헌 조사를 통해 분석하였다. 일차측 환경에서 주기적인 반복하중을 받을 때에는 기계적인 피로균열에 더해 수소유기균열이나 동적변형시효 등으로 인한 가속화 메커니즘을 통해 피로수명 감소가 나타났다. 따라서 EAF에 대한 저항성은 전반적인 부식저항성이 우수한 니켈기합금이 스테인리스강보다 크게 나타났다. 그러나 일정한 하중을 받을 때에는 내부산화에 의해 국부적인 취약부인 입계로의 빠른 균열의 생성과 진전이 나타나 일차수 응력부식균열(PWSCC)이라는 형태로 발생한다고 여겨진다. 이때는 니켈-크롬의 비율이 내부산화 저항성에 영향을 미쳐, 비율이 낮은 스테인리스강은 높은 저항성을 가지고, 비율이 높은 니켈기합금은 낮은 저항성을 가진다. 그러나 아직 이러한 균열 메커니즘에 대한 명확한 이해가 부족하므로, 명확히 규명하기 위해서는 추가적인 연구가 필요하다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.83-84
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• 2008

The primary water stress corrosion cracking (PWSCC) of dissimilar metal weld based on Alloy 82/182 is one of major issues in material degradation of nuclear components. It is well known that the crack initiation and growth due to PWSCC is influenced by material's susceptibility to PWSCC and distribution of welding residual stress. Therefore, modeling the welding residual stress is of interest in understanding crack formation and growth in dissimilar metal weld. Currently in Korea, a numerical round robin study is undertaken to provide guidance on the welding residual stress analysis of dissimilar metal weld. As a part of this effort, the present paper investigates distribution of welding resisual stress of a ferritic low alloy steel nozzle with dissimilar metal weld using Alloy 82/182. Two-dimensional thermo-mechanical finite element analyses are carried out to simulate multi-pass welding process on the basis of the detailed design and fabrication data. The present results are compared with those from other participants, and more works incorporating physical measurements are going to be performed to quantify the uncertainties relating to modelling assumptions.

• 한국압력기기공학회 논문집
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• 제6권2호
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.